Recent Progress in the NSTX/NSTX-U Lithium Program and Prospects for Reactor-Relevant Liquid-Lithium Based Divertor Development
Authors: M. Ono, et al.
Abstract: Developing a reactor compatible divertor has been identified as a particularly challenging technology
problem for magnetic confinement fusion. While tungsten has been identified as the most attractive solid
divertor material, the NSTX/NSTX-U lithium (Li) program is investigating the viability of liquid lithium (LL) as
a potential reactor compatible divertor plasma facing component (PFC) . In the near term, operation in NSTX-U
is projected to provide reactor-like divertor heat loads < 40 MW/m^2 for 5 s. During the most recent NSTX
campaign, ~ 0.85 kg of Li was evaporated onto the NSTX PFCs where a ~50% reduction in heat load on the
Liquid Lithium Divertor (LLD) was observed, attributable to enhanced divertor bolometric radiation. This
reduced divertor heat flux through radiation observed in the NSTX LLD experiment is consistent with the results
from other lithium experiments and calculations. These results motivate an LL-based closed radiative divertor
concept proposed here for NSTX-U and fusion reactors. With an LL coating, the Li is evaporated from the
divertor strike point surface due to the intense heat. The evaporated Li is readily ionized by the plasma due to
its low ionization energies, and the ionized Li ions can radiate strongly, resulting in a significant reduction in the
divertor heat flux. Due to the rapid plasma transport in divertor plasma, the radiation values can be significantly
enhanced up to ~ 11 MJ/cc of LL. This radiative process has the desired function of spreading the focused
divertor heat load to the entire divertor chamber facilitating the divertor heat removal. The LL divertor surface
can also provide a "sacrificial" surface to protect the substrate solid material from transient high heat flux such
as the ones caused by the ELMs. The closed radiative LLD concept has the advantages of providing some degree
of partition in terms of plasma disruption forces on the LL, Li particle divertor retention, and strong divertor
pumping action from the Li-coated divertor chamber wall. By operating at a lower temperature than the first
wall, the LLD can serve to purify the entire reactor chamber, as impurities generally migrate toward lower
temperature Li-condensed surfaces. To maintain the LL purity, a closed LL loop system with a modest capacity
(e.g., ~ 1 Liter/sec for ~ 1% level "impurities") is envisioned for a steady-state 1 GW-electric class fusion power
plant.
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Submitted to: IAEA Oriceedubgsm (September 2012)
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Download PPPL-4830 (pdf KB pp)
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